System for recovery of daughter isotopes from a source material

ABSTRACT

A method of separating isotopes from a mixture containing at least two isotopes in a solution is disclosed. A first isotope is precipitated and is collected from the solution. A daughter isotope is generated and collected from the first isotope. The invention includes a method of producing an actinium-225/bismuth-213 product from a material containing thorium-229 and thorium-232. A solution is formed containing nitric acid and the material containing thorium-229 and thorium-232, and iodate is added to form a thorium iodate precipitate. A supernatant is separated from the thorium iodate precipitate and a second volume of nitric acid is added to the thorium iodate precipitate. The thorium iodate precipitate is stored and a decay product comprising actinium-225 and bismuth-213 is generated in the second volume of nitric acid, which is then separated from the thorium iodate precipitate, filtered, and treated using at least one chromatographic procedure. A system for producing an actinium-225/bismuth-213 product is also disclosed.

CROSS-REFERENCE TO RELATED APPLICATION

This application is a divisional of U.S. application Ser. No.10/247,016, filed Sep. 18, 2002, now U.S. Pat. No. 6,951,634, issuedOct. 4, 2005.

GOVERNMENT RIGHTS

The United States Government has certain rights in this inventionpursuant to Contract No. DE-AC07-991D13727, and Contract No.DE-AC07-051D14517 between the United States Department of Energy andBattelle Energy Alliance, LLC.

TECHNICAL FIELD

The invention pertains to methods of producing isotopes from a sourcematerial and a system for producing an actinium-225/bismuth-213 product.

BACKGROUND OF THE INVENTION

Radioisotopes are becoming increasingly useful in various scientific andmedical fields. Many radioisotopes are currently used for researchtechniques such as radiolabeling for detection and monitoring purposes.In the medical field, numerous radionuclides are used for a variety ofdiagnostic and treatment techniques.

New immunotherapy techniques are currently being developed for treatmentof various forms of cancer. Such techniques can include using a nuclideto label an antibody targeted to a tumor, thereby utilizing the labeledantibody to deliver the nuclide to the tumor site. The emissionsgenerated by radioactive decay of the nuclide label can thereby be moreselectively localized to the tumor site relative to other methods ofradiotherapy.

In many techniques that utilize radionuclides, it can be desirable toutilize nuclides capable of generating alpha emissions. For example,nuclides that emit alpha particles of relatively high energy can bedesirable for immunotherapy applications to maximize localization ofdecay emissions for the selective destruction of tumor cells whileminimizing damage to surrounding tissues.

Although radioisotopes are becoming increasingly useful, it is oftendesirable to maximize purity of the nuclide prior to its ultimate use.However, desired radionuclides are often generated in very smallquantities within other materials from which the nuclide must beseparated. It is therefore desirable to develop methods of separatingisotopes from source materials.

SUMMARY OF THE INVENTION

In one aspect, the invention encompasses a method of separating isotopesfrom a mixture. A mixture containing at least two isotopes is used toform a solution. A precipitate containing a first isotope is formed andis collected from the solution. One or more daughter isotopes aregenerated from the first isotope and at least one of the one or moredaughter isotopes is collected.

In one aspect, the invention encompasses a method of producing anactinium-225/bismuth-213 product from a thorium source material. Athorium source material containing thorium-229 and thorium-232 isprovided. A solution is formed containing a first volume of nitric acidand at least some of the thorium source material. Iodate is added to thesolution and at least some of the iodate combines with thorium-229 andthorium-232 to form a thorium iodate precipitate. A supernatantcontaining at least some of the first volume of nitric acid is separatedfrom the thorium iodate precipitate and a second volume of nitric acidis added to the thorium iodate precipitate. The thorium iodateprecipitate is stored in the second volume of nitric acid for ageneration time period during which a thorium-229 decay productcomprising actinium-225 and bismuth-213 is generated. The second volumeof nitric acid containing at least some of the thorium-229 decay productis separated from the thorium iodate precipitate and is filtered toremove at least some of any residual thorium iodate precipitate present.After filtering, the second volume of nitric acid is treated using atleast one chromatographic procedure to separate actinium-225 andbismuth-213 from at least some of any impurities that are present in thesecond volume of nitric acid.

In another aspect, the invention includes a system for producing anactinium-225/bismuth-213 product.

BRIEF DESCRIPTION OF THE DRAWINGS

Preferred embodiments of the invention are described below withreference to the following accompanying drawings.

FIG. 1 is a block diagram flow chart view of a method encompassed by thepresent invention.

FIG. 2 is a block diagram illustration of the uranium-233 decay chain.The half-lives of radioisotopes are indicated in years (y), days (d),hours (h), minutes (m), seconds (s), or milliseconds (ms).

FIG. 3 is a graph depicting the generation of actinium-225 during aprocessing step of methodology of the present invention.

FIG. 4 is a schematic view showing an exemplary production system thatcan be utilized in performing methods of the present invention.

DETAILED DESCRIPTION OF THE PREFERRED EMBODIMENTS

The invention encompasses methodology for separating isotopes from amixture and also encompasses a system for producing isotopes utilizingmethods of the present invention. A process encompassed by the presentinvention is described generally with reference to the block diagram ofFIG. 1. In an isotope production flow scheme 10, a material comprising aparent isotope is provided in an initial act 12. The material providedin act 12 can comprise two or more isotopes, at least one of theisotopes present in the material being a parent isotope. For purposes ofthe present description, use of the term “parent isotope” can refer toan isotope capable of undergoing decay to produce a daughter isotope.Further decay of a daughter isotope can result in a “granddaughter”isotope of the original parent isotope. For purposes of the presentdescription, the term “daughter isotope” can refer to a first decayproduct isotope generated from a parent or any subsequent decay productisotope generated from the first decay product isotope.

The material provided in act 12 can comprise any source materialcontaining a parent isotope capable of generating a desired daughterisotope. In particular embodiments, the material provided in act 12 willcontain a parent isotope capable of generating a daughter isotope thatcan further undergo alpha decay. In particular embodiments, the materialprovided in act 12 can comprise thorium-229. Alternatively, the materialprovided in act 12 can comprise thorium-228, or can comprise boththorium-228 and thorium-229.

In embodiments of the present invention where the material provided inact 12 comprises thorium-229, the thorium-229 can function as a parentisotope. Referring to FIG. 2, which shows the decay chain ofuranium-233, it is noted that thorium-229 is itself a decay productisotope resulting from decay of uranium-233. Thorium-229 can undergo aseries of decays, as shown in FIG. 2, to produce thorium-daughtersincluding actinium-225. Actinium-225 can further undergo a series ofdecays to generate daughters that include bismuth-213. As shown in FIG.2, bismuth-213 is capable of decay by alpha emission. Additionally,bismuth-213 has a half-life of forty-six minutes and undergoes a seriesof decays to form a stable isotope bismuth-209. Accordingly, athorium-229 comprising source material can be utilized in act 12 of FIG.1 for generation of one or both of actinium-225 and bismuth-213 bymethods of the present invention.

A thorium-229 source material in act 12 can comprise additional thoriumisotopes such as, for example, thorium-232. The methods of the presentinvention do not require a specific purity of thorium-229 and canutilize materials comprising any specific activity ofthorium-229/thorium-2XX. An exemplary thorium source material cancomprise a thorium-229/thorium-2XX ratio of as little as 50 grams ofthorium-229 to about 14 metric tons of thorium-232. Alternatively, ahigher purity of thorium-229 can be utilized but is not required formethods of the present invention.

A thorium-229 source material can comprise additional non-thoriumisotopes such as, for example, uranium-233. A uranium/thorium (U/Th)source material can comprise any uranium to thorium ratio. An exemplarymaterial for utilization in act 12 can comprise an unirradiated U/Thnuclear fuel, such as U/Th oxide fuel pellets. U/Th fuel pellets can beany available pellet size such as, for example, pellets comprising alength of about 0.5 inch and a diameter of about 0.5 inch.

The material provided in act 12 can be processed to form a solution inact 14. Solution formation of act 14 can comprise mixing the materialcomprising a parent isotope from act 12 with a solvent such as an acidsolution, for example, a nitric acid solution. A nitric acid solutionuseful for dissolving the parent isotope material can comprise fromabout 8 M nitric acid to about 14.9 M nitric acid, and preferablycomprises about 13.6 M nitric acid.

Solution formation of act 14 can further comprise addition of a catalystto enhance the dissolving. For example, the dissolving can be catalyzedby addition of from 0.01 M hydrofluoric acid to about 0.05 Mhydrofluoric acid. Additionally, the solution formation of act 14 cancomprise either intermittent or continuous stirring to assist thedissolving of the parent isotope source material. Solution formation ofact 14 can comprise a temperature from ambient temperature to about 300°C. If a temperature above ambient temperature is utilized, the increasein temperature above ambient can be accompanied by a correspondingincrease in pressure.

Preferably, the parent isotope material is fully dissolved duringsolution formation of act 14. Alternatively, the solution can betreated, for example, by filtering, to remove non-dissolved materialprior to subsequent processing.

The solution comprising parent isotope source material can be treated toselectively precipitate one or more isotopes from the solution inprecipitation of act 16. Precipitate formation of act 16 preferablyproduces a precipitated material comprising at least one desired parentisotope. The precipitation can comprise using one or more ion speciescapable of combining with the parent isotope to form a precipitate. Theone or more ion species can be provided by, for example, addition of oneor more salts to the solution. Salt can be added in solid form or in astock solution.

In embodiments of the present invention where a thorium source materialis utilized, thorium isotopes including thorium-229 parent isotope canbe precipitated from the solution formed in act 14 by addition of, forexample, iodate ion in act 16. Preferably, the amount of iodate added toa solution comprising thorium will exceed the total amount of thoriumisotopes dissolved in the solution. It can be advantageous to add anexcess stoichiometric ratio of iodate relative to the totalconcentration of all thorium isotopes present in the tank to maximizethorium iodate precipitation.

Providing iodate to the thorium solution in act 16 can additionallycomprise formation of a stock iodate solution which can be formed bycombining nitric acid with one or more iodate salts selected from thegroup consisting of KIO₃, NaIO₃, or HIO₃. An exemplary stock iodatesolution can comprise 1 M iodate and 6 N nitric acid. It can beadvantageous to form the stock iodate solution utilizing the acid iodatesalt to avoid introducing counter ions that can potentially interferewith any downstream chemistry. It is to be noted that the iodate stocksolution concentration is exemplary and any concentration of stocksolution can be utilized for precipitation of act 16.

A precipitation collection of act 18 can be performed to collect theprecipitate formed in act 16. Precipitate collection of act 18 cancomprise, for example, removal of the supernatant to separate theprecipitate from materials that remain dissolved in the solution. Inembodiments of the present invention that utilize a U/Th parent isotopesource material in act 12, precipitation collection of act 18 cancomprise separation of a precipitated thorium material, such as thoriumiodate, from a supernatant comprising uranium. Supernatant removal cancomprise a filtration process such as, for example, cross-flowfiltration. In embodiments wherein the recovered supernatant comprisesuranium, such supernatant can be sent to a uranium storage unit whereany soluble thorium daughters present in the supernatant will eventuallydecay. After storage of the recycled uranium for a length of time toproduce a sufficient amount of thorium-229, the stored material can bereprocessed utilizing methods of the present invention.

Precipitate collection of act 18 can optionally include a precipitatewash step to minimize or eliminate any residual uranium present afterremoval of the initial supernatant. The precipitate wash step cancomprise one or more rounds of a wash solution, followed by removal ofthe wash solution. Removal of the wash solution can comprise, forexample, filtration including, but not limited to, cross-flowfiltration.

In embodiments of the present invention wherein the precipitatecollected in act 18 comprises thorium, an appropriate wash solution cancomprise, for example, a salt and acid solution, preferably a nitricacid iodate solution. It can be advantageous to utilize an iodatesolution for washing a thorium precipitate collected in act 18 tominimize or eliminate any dissolving of the thorium precipitate duringthe washing. An exemplary wash solution for washing a thorium iodateprecipitate can comprise about 1 M iodate and from about 4 N nitric acidto about 7 N nitric acid. An exemplary washing step for washing athorium precipitate can comprise, for example, three rounds of washing,wherein each washing step adds a volume of a nitric acid iodatesolution, followed by removal of the volume of wash solution by, forexample, cross-flow filtration. The washing of thorium iodateprecipitate by methods of the present invention can minimize anyresidual uranium remaining in the collected precipitate.

Although isotope production flow scheme 10 shows a single series of acts12-18, it is to be understood that additional rounds of providing aparent isotope material of act 12, solution formation of act 14,precipitation of act 16, and collection of the precipitate of act 18 canbe performed, and the resulting precipitates can be combined to create adesired amount of precipitate for further processing. The collectedprecipitates can be combined either prior to, or subsequent to washingof precipitates. The resulting combined precipitates can then beutilized for further processing according to methods of the presentinvention.

The precipitate collected in act 18 can be utilized to generate decayproducts from a parent isotope in act 20. The generation of decayproducts in act 20 can comprise storing a precipitate from act 18 for astorage period of a sufficient time length to generate a decay product,also called an “in-growth” period. It is possible to determine anappropriate in-growth period for generation of a daughter isotope basedupon the activity of the parent isotope and the half-life of thedaughter isotope. For example, referring to FIG. 3 which shows anin-growth curve of actinium-225 as a percent of thorium-229 activityover time, it can be seen that the amount of actinium-225 that can begenerated and collected from a thorium-229 comprising precipitatereaches a maximum in-growth level by approximately 100 days.

Although the recoverable amount of actinium cannot be increased byin-growth periods of longer than approximately 100 days, in someinstances it may be advantageous to store thorium-229 comprisingprecipitate for longer than a 100 day period based on product demand.For example, it may be advantageous to delay collection of a decayproduct comprising actinium-225 generated from a thorium-229 comprisingprecipitate to minimize the amount of actinium-225 decay that occursbetween decay product collection and any subsequent delivery or use ofan actinium-225 or actinium-225/bismuth-213 (Ac-225/Bi-213) product(discussed below). Accordingly, a thorium-229 comprising precipitate canbe incubated to allow in-growth of actinium-225 for a period of zero togreater than 100 days. Preferably, the in-growth period can be from tenand 100 days and in particular embodiments can comprise an amount oftime determined by product demand.

A storage solution can be utilized for storage of the precipitate duringthe in-growth incubation period and can preferably provide selectivesolubilization of at least one decay product relative to its parentisotope. Storage of act 20 preferably comprises storage of theprecipitate in a storage solution capable of minimizing any dissolvingof the precipitate. Preferably, the parent isotope precipitate willcomprise a very low solubility in the storage solution and at least onedaughter isotope generated from the parent isotope will comprise anincreased solubility in the storage solution relative to the precipitatecomprising the parent isotope.

An exemplary storage solution for storage of collected precipitatescomprising parent isotope thorium-229 can comprise nitric acid andiodate. The nitric acid iodate storage solution can comprise, forexample, the concentrations of nitric acid and iodate discussed abovewith respect to the wash solution. Thorium iodate has a very lowsolubility in nitric acid solutions and can comprise a solubility(k_(s)) of about 2.5×10⁻¹⁵ in the presence of a stoichiometric excess ofiodate relative to thorium. Daughter isotopes that result from decay ofthorium-229 such as, for example, actinium-225 and bismuth-213, have anincreased solubility in a storage solution comprising nitric acid andiodate relative to the thorium iodate precipitate. Accordingly, a decayproduct comprising one or both of actinium-225 and bismuth-213(hereafter referred to as an Ac-225/Bi-213 decay product) can begenerated from parent isotope thorium-229 and the resultingAc-225/Bi-213 decay product can be dissolved by the storage solution.

Migration or diffusion of decay products out of a crystal precipitateduring the in-growth storage can be enhanced by agitation of theprecipitate. The agitation can preferably be constant agitation and canutilize, for example, mixing, stirring, air sparging or sparging with aninert gas.

After an appropriate in-growth storage period for generation of decayproducts in act 20, a decay product collection of act 22 can beperformed. Collection of a decay product in act 22 can comprise removingdissolved decay product from the parent isotope by separating thestorage solution as a liquid fraction from the precipitate comprisingthe parent isotope. For example, in embodiments wherein the precipitatecomprises parent isotope thorium-229, an Ac-225/Bi-213 decay product canbe recovered in a liquid fraction comprising the storage solution sincethe actinium and bismuth daughter isotopes have higher solubilities inthe storage solution than does the thorium iodate precipitate.

Collection of a decay product in act 22 can comprise separation of aliquid fraction from a solid fraction by, for example, utilization ofone or more filtration techniques including, but not limited to,cross-flow filtration. Upon removal, the filtrate can be recycled foradditional use as a storage solution. After removal of the liquidfraction, a solid portion comprising a parent isotope can be used togenerate additional quantities of decay product by repeated rounds ofsequentially adding storage solution to the precipitate, and incubatingthe recovered precipitate for an additional in-growth period to generateadditional decay product.

Decay product collection of act 22 can further comprise a purificationprocess involving one or more purification techniques designed tominimize any residual parent isotope or other impurity present in thefraction comprising the decay product. Purification techniques that canbe utilized in the decay product collection of act 22 include, forexample, one or both of extraction chromatography techniques and columnchromatography techniques, including but not limited to, ion exchangechromatography. It can be advantageous to use chromatographic techniquesthat are compatible with the storage solution to avoid having to changesolvents prior to the chromatographic step. It is to be understood,however, that the present invention contemplates chromatographytechniques utilizing alternative, non-storage solvent systems.

In particular embodiments of the present invention, purificationtreatment can comprise a series of column chromatography steps utilizingat least two different chromatographic separation techniques. Forexample, a first column chromatography step can be specifically designedto separate daughter isotopes comprised by the decay product from aresidual parent isotope. A second column chromatography act can bespecifically configured to separate any non-parent isotope impuritiespresent in the storage solution from the daughter isotope product.Alternatively, the exemplary column chromatography techniques can beperformed in the reverse order. Additional chromatography techniques canbe utilized to maximize purity of the decay product collected in act 22.

The decay product collected in act 22 can be contained in a vessel priorto its ultimate use. Preferably, the vessel is inert to the collectedproducts and any additional products produced by further decay.Additionally, the vessel is preferably inert to a final solventcontaining the collected decay product. When the end use of the productis at a location remote from the site of decay product generation, thevessel can further serve as a shipping vessel for delivery of the decayproduct to the site of its end use.

In particular embodiments of the present invention, the collected decayproduct can be retained on a chromatography column which can serve as acontainment vessel until the decay product end use. Further, a columncomprising the retained decay product can be utilized as a shippingvessel to deliver the product to a desired use site. In embodimentshaving the decay product retained on a column, the decay product can beeluted or stripped from the column at the use site.

Elution of the decay product from the final chromatography step involvedin product purification can comprise elution into a final eluentdesigned to be compatible with a particular end use. For example, inembodiments wherein the product will be utilized for medical purposes,the decay product can be eluted from a final purification columnutilizing a elution system that is compatible with a medical use.Alternatively, the eluted product can be dried, shipped, andre-dissolved in a use-compatible solvent at the use site.

The chromatographic steps utilized in decay product purification canfurther serve to concentrate the decay products into a final volume.Alternatively, an additional concentration step can be performed toconcentrate the decay product.

An exemplary system which can be utilized for performing methodsencompassed by the present invention is described with reference to FIG.4. An isotope decay product production system 100 is shown. Isotopedecay product production system 100 can be used, for example, forproduction of an Ac-225/Bi-213 decay product generated from a thoriumcomprising source material. For purposes of the present description,production system 100 can be referred to as an Ac-225/Bi-213 productionsystem. Although production system 100 is described as an Ac-225/Bi-213production system, it is to be understood that the present inventioncontemplates use of production system 100 for production of other decayproducts including products generated from non-thorium source material.

Ac-225/Bi-213 production system 100 can comprise an isotope source 101configured to provide an isotope source material, for example, athorium-229 source material, into a dissolver unit 102. A solution canbe formed in dissolver unit 102 by providing a solvent comprising, forexample, 13.6 N nitric acid. Dissolving thorium-229 source material indissolver unit 102 can be catalyzed by, for example, providing fromabout 0.01 N hydrofluoric acid to about 0.05 N hydrofluoric acid intodissolver unit 102.

Dissolver unit 102 can be configured to stir the mixture during thedissolving. Further, the dissolving can comprise a temperature fromabout ambient temperature to about 300° C.

After dissolving at least some of the thorium-229 source material toform a solution, the solution can be fed into a precipitation tank 104.The solution can be fed into precipitation tank 104 continuously oralternatively can be provided batch wise. Preferably, the solution isprovided into the precipitation tank 104 batch wise.

A thorium precipitate can be formed within precipitation tank 104 by,for example, providing a salt source 106 configured to provide a salt tothe precipitation tank 104. Salt source 106 can comprise, for example, asalt solution containing an iodate salt stock solution comprising nitricacid, as discussed above. Iodate provided to precipitation tank 104 cancombine with dissolved thorium isotopes to form a thorium iodateprecipitate. The thorium iodate precipitate can comprise any of thethorium isotopes contained in the thorium source material including, butnot limited to, thorium-229 and thorium-232. It is advantageous toprovide a concentration of iodate to precipitation tank 104 thatstoichiometrically exceeds a total concentration of thorium isotopeswithin the precipitation tank 104 to maximize and maintain precipitationof thorium iodate.

After formation of the thorium iodate precipitate within precipitationtank 104, a supernatant can be removed to separate at least some of anynon-precipitated materials from the thorium iodate precipitate. Inembodiments where the supernatant contains dissolved uranium, thesupernatant can be provided to a uranium storage unit 105. Thesupernatant can be removed from precipitation tank 104 by, for example,draining off the supernatant or by filtration, for example, bycross-flow filtration (not shown).

After removal of the supernatant, the thorium iodate precipitate can bewashed to minimize any residual uranium or any other soluble impuritiesby, for example, repeated rounds of sequentially adding a volume of awash solution to precipitation tank 104 and subsequent removal of thewash solution from the precipitation tank 104. The acid iodate washsolution can be formed, for example, as discussed above. It can beadvantageous to wash the thorium precipitate a number of times, forexample, using three rounds of washing, to maximize removal of anyresidual uranium or other soluble impurities.

The resulting thorium iodate precipitate can be combined with otherthorium iodate precipitates produced by methods of the present inventionprior to subsequent processing. After the removal of the supernatant andoptional wash steps, a volume of storage solution comprising nitric acidand iodate can be added to the thorium iodate precipitate inprecipitation tank 104. The thorium iodate precipitate can be stored inprecipitation tank 104 for an in-growth period sufficient to generate atleast some decay product. Preferably, the thorium iodate precipitate isstored in precipitation tank 104 for an in-growth period of from aboutten to about 100 days and more preferably for about 30 days. A preferredtemperature for storage of the thorium iodate precipitate can be fromabout 5° C. to about 30° C. It is to be understood that higher storagetemperatures are contemplated. During the in-growth period, the thoriumiodate precipitate in storage solution can be agitated by mixing orsparging to enhance diffusion of daughter isotopes, such as actinium-225and bismuth-213 out of the crystal thorium iodate precipitate.

Upon expiration of the in-growth time period, a liquid fractioncontaining a decay product can be removed from the precipitation tank104. Removal of a liquid fraction from precipitation tank 104 cancomprise utilization of a separator 108. Separator 108 can comprise, forexample, a filtration system, such as, for instance, a cross-flowfiltration system. In embodiments of the present invention whereseparator 108 comprises cross-flow filtration, the cross-flow filter canbe configured such that dissolved thorium daughters will pass through aninner membrane while any thorium iodate precipitate present can beretained within an inner diameter of the cross-flow filter. Thecross-flow filtration system can comprise an internal pressure gradientacross a membrane of from about 20 lbs to about 500 lbs. The thoriumdaughters can thereby be recovered from the cross-flow filtration systemand can be further purified as discussed below.

Ac-225/Bi-213 production system 100 can be further configured such thatany thorium iodate precipitate that is retained within an inner diameterof a cross-flow filter of separator 108 can be recycled back toprecipitation tank 104 as indicated by recycle path 110. The thoriumiodate precipitate remaining in precipitation tank 104 can be utilizedfor a subsequent round of thorium decay product generation.

The thorium decay product collected from separator 108 can comprise atleast one of actinium-225 and bismuth-213 (Ac-225/Bi-213), preferablythe decay product comprises both actinium-225 and bismuth-213. Aftercollection from separator 108, the Ac-225/Bi-213 decay product can befurther purified utilizing a first purifier 112. First purifier 112 cancomprise, for example, an anion exchange column comprising, forinstance, DOWEX® 1×8 resin (Dow Corning Corp., Michigan, U.S.A.).Alternatively, other extraction chromatography methods or columnchromatography techniques, including alternate anion exchangetechniques, can be utilized for selectively separating thorium from itssoluble daughter isotopes. It can be advantageous to use DOWEX® 1×8resin for separating thorium from Ac-225/Bi-213 decay product sinceDOWEX® 1×8 resin is compatible with the nitric acid storage solvent andcan avoid an added step of changing solvent prior to first purifier 112.

Ac-225/Bi-213 production system 100 can be further configured to providea recycle pathway 114 to deliver thorium recovered from first purifier112 back to precipitation tank 104.

The Ac-225/Bi-213 decay product recovered from first purifier 112 can beprovided to a second purifier 116. Second purifier 116 can comprise, forexample, a cation exchange column. Alternatively, second purifier 116can comprise extraction chromatography or other column chromatographytechniques capable of selective separation of actinium and actiniumdaughters from other materials present in the nitric acid storagesolution. Ac-225/Bi-213 product can be selectively eluted or strippedfrom the cation exchange column of second purifier 116 and can beprovided to a collector 120. Ac-225/Bi-213 production system 100 canfurther comprise a recycle pathway 118 configured to deliver nitric acidand/or iodate recovered from second purifier 116 back to precipitationtank 104.

Although FIG. 4 shows two purifiers, it is to be understood thatAc-225/Bi-213 production system 100 can comprise a single purifier ormore than two purifiers (not shown). Preferably, Ac-225/Bi-213production system 100 comprises sufficient purifiers to produce adesired decay product purity. In particular embodiments of the presentinvention where the Ac-225/Bi-213 product is intended for medical use,Ac-225/Bi-213 production system 110 will comprise a sufficient number ofpurifiers for production of a medical grade Ac-225/Bi-213 decay product.

In particular embodiments of the present invention, the decay productproduced by production system 100 will be utilized as a source of anisotope capable of alpha emission. For example, an Ac-225/Bi-213 decayproduct can be used as a source for generation of bismuth-213, which iscapable of decay by alpha emission as shown in FIG. 2. In particularembodiments, collector 120 can comprise a containment vessel configuredfor delivery of the decay product to a user. When a decay productproduced by production system 100 comprises an Ac-225/Bi-213 product,the containment vessel can preferably comprise a material that is inertto actinium-225 and the decay products of actinium-225, and also inertto a solvent used to elute or strip the decay product from secondpurifier 116.

In particular embodiments of the present invention, second purifier 116can be utilized as the containment vessel. For example, an Ac-225/Bi-213decay product produced by production system 100 can be retained withinsecond purifier 116 and can be eluted just prior to use.

An Ac-225/Bi-213 product produced according to methods of the presentinvention can be used, for example, as a source of bismuth-213 forgeneration of alpha decay emissions. Bismuth-213 can decay by producingalpha emissions having an extremely high energy of about 8.4 MeV. Due toits ability to produce high energy alpha emissions, bismuth-213 can beespecially desirable for use as a nuclide for use in immunotherapyapplications. Accordingly, the present invention can be utilized forproduction of bismuth-213 for uses such as research, medical diagnosticsand medical treatments, including immunotherapy.

Although the embodiments discussed above refer to production ofactinium-225 and/or bismuth-213, the invention can be utilized forproduction of radium-224 and/or its decay product bismuth-212.Thorium-228, which is a parent isotope of radium-224, occurs in very lowquantities in natural thorium. The methods discussed above can beutilized to precipitate thorium-228 and to recover soluble daughterstherefrom. Further, the separation and purification methods discussedabove can be adapted to selectively separate a Ra-224/Bi-212 productfrom a thorium-228 source material.

In compliance with the statute, the invention has been described inlanguage more or less specific as to structural and methodical features.It is to be understood, however, that the invention is not limited tothe specific features shown and described, since the means hereindisclosed comprise preferred forms of putting the invention into effect.The invention is, therefore, claimed in any of its forms ormodifications within the proper scope of the appended claimsappropriately interpreted in accordance with the doctrine ofequivalents.

1. A system for producing bismuth-213, comprising: a dissolver unit; aprecipitation tank configured to receive a dissolved thorium materialfrom the dissolver unit; a salt source configured to deliver one or moresalt species to the precipitation tank; a separator configured toreceive a liquid fraction comprising a thorium decay product includingat least one of actinium-225 and bismuth-213 from the precipitationtank; at least one chromatography column configured to receive thethorium decay product including at least one of actinium-225 andbismuth-213 from the separator; a collector configured to receive andretain at least one of actinium-225 and bismuth-213 from the at leastone chromatography column; at least one recycle pathway configured todeliver thorium recovered from the at least one chromatography column tothe precipitation tank; and a pathway for delivering a supernatant fromthe precipitation tank to a storage.
 2. The system of claim 1, whereinthe at least one chromatography column comprises a plurality ofchromatography columns in series, the at least one recycle pathwaycomprises a recycle pathway from each of the plurality of chromatographycolumns to the precipitation tank, and the collector is configured toreceive the at least one of actinium-225 and bismuth-213 from a last inthe series of the plurality of chromatography columns.
 3. The system ofclaim 1, wherein the separator is a cross-flow filtration device.
 4. Thesystem of claim 1, wherein the salt source is an iodate salt source. 5.The system of claim 1, further comprising: a source of thorium material,wherein the thorium material comprises thorium and uranium; and whereinthe storage comprises a uranium storage and wherein the supernatantcomprises dissolved uranium.
 6. The system of claim 1, wherein theseparator is configured to separate at least some of any thorium presentin the liquid fraction from the at least one of actinium-225 andbismuth-213 to produce a thorium recovery fraction, and wherein thethorium recovery fraction is recycled back into the precipitation tank.7. A system for producing bismuth-212, comprising: a dissolver unit; aprecipitation tank configured to receive a dissolved thorium materialfrom the dissolver unit; a salt source configured to deliver one or moresalt species to the precipitation tank; a separator configured toreceive a liquid fraction comprising a thorium decay product includingat least one of radium-224 and bismuth-212 from the precipitation tank;at least one chromatography column configured to receive the thoriumdecay product including at least one of radium-224 and bismuth-212 fromthe separator; a collector configured to receive and retain at least oneof radium-224 and bismuth-212 from the at least one chromatographycolumn; at least one recycle pathway configured to deliver thoriumrecovered from the at least one chromatography column to theprecipitation tank and a pathway for delivering a supernatant from theprecipitation tank to a storage.
 8. The system of claim 7, wherein theat least one chromatography column comprises a plurality ofchromatography columns in series, the at least one recycle pathwaycomprises a recycle pathway from each of the plurality of chromatographycolumns to the precipitation tank, and the collector is configured toreceive the at least one of radium-224 and bismuth-212 from a last inthe series of the plurality of chromatography columns.
 9. The system ofclaim 7, wherein the salt source is an iodate salt source.
 10. Thesystem of claim 7, wherein the separator is a cross-flow filtrationdevice.
 11. The system of claim 7 further comprising: a source ofthorium material, wherein the thorium material comprises thorium anduranium; and wherein the storage comprises a uranium storage and whereinthe supernatant comprises dissolved uranium.
 12. The system of claim 7,wherein the separator is configured to separate at least some of anythorium present in the liquid fraction from the at least one ofradium-224 and bismuth-212 to produce a thorium recovery fraction, andwherein the thorium recovery fraction is recycled back into theprecipitation tank.